The National Spherical Torus Experiment (NSTX) Research Program andProgress Towards High Beta, Long Pulse Operating Scenarios

E.J. Synakowski1, M.G. Bell1, R.E. Bell1, T. Bigelow2, M. Bitter1, W. Blanchard1, J. Boedo3,C. Bourdelle4, C. Bush2, D.S. Darrow1, P.C. Efthimion1, E.D. Fredrickson1, D.A. Gates1, M.Gilmore5, L.R. Grisham1, J.C. Hosea1, D.W. Johnson1, R. Kaita1, S.M. Kaye1, S. Kubota6,H.W. Kugel1, B.P. LeBlanc1, K. Lee7, R. Maingi2, J. Manickam1, R. Maqueda8, E.Mazzucato1, S.S. Medley1, J. Menard1, D. Mueller1, B.A. Nelson9, C. Neumeyer1, M. Ono1, F.Paoletti10, H.K. Park1, S.F. Paul1, Y.-K. M. Peng2, C.K. Phillips1, S. Ramakrishnan1, R.Raman9, A.L. Roquemore1, A. Rosenberg1, P.M. Ryan2, S.A. Sabbagh10, C.H. Skinner1, V.Soukhanovskii1, T. Stevenson1, D. Stutman11, D.W. Swain2, G. Taylor1, A. Von Halle1, J.Wilgen2, M. Williams1, J.R. Wilson1, X. Xu12, S.J. Zweben1, R. Akers13, R.E. Barry2, P.Beiersdorfer12, J.M. Bialek10, B. Blagojevic11, P.T. Bonoli14, R. Budny1, M.D. Carter2, J.Chrzanowski1, W. Davis1, B. Deng7, E.J. Doyle15, L. Dudek1, J. Egedal14, R. Ellis1, J.R.Ferron15, M. Finkenthal11, J. Foley1, E. Fredd1, A. Glasser8, T. Gibney1, R.J. Goldston1, R.Harvey16, R.E. Hatcher1, R.J. Hawryluk1, W. Heidbrink17, K.W. Hill1, W. Houlberg2, T.R.Jarboe9, S.C. Jardin1, H. Ji1, M. Kalish1, J. Lawrance18, L.L. Lao15, K.C. Lee7, F.M.Levinton19, N.C. Luhmann7, R. Majeski1, R. Marsala1, D. Mastravito1, T.K. Mau3, B.McCormack1, M.M. Menon2, O. Mitarai20, M. Nagata21, N. Nishino22, M. Okabayashi1, G.Oliaro1, D. Pacella23, R. Parsells1, T. Peebles6, B. Peneflor15, D. Piglowski15, R. Pinsker15,G.D. Porter12, A.K. Ram14, M. Redi1, M. Rensink12, G. Rewoldt1, J. Robinson1, P. Roney1, M.Schaffer15, K. Shaing24, S. Shiraiwa25, P. Sichta1, D. Stotler1, B.C. Stratton1, Y. Takase25, X.Tang8, R. Vero11, W.R. Wampler26, G.A. Wurden8, X.Q. Xu12, J.G. Yang27, L. Zeng6, W.Zhu6

First author e-mail address: synakowski@pppl.gov

1 Princeton Plasma Physics Laboratory, Princeton University, Princeton, New Jersey,United States of America2 Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States of America3 University of California, San Diego, California, United States of America4 CEA Cadarache, France5 University of New Mexico at Albuquerque6 University of California, Los Angeles, California, United States of America7 University of California, Davis, California, United States of America8 Los Alamos National Laboratory, Los Alamos, New Mexico, United States of America9 University of Washington, Seattle, Washington, United States of America10 Columbia University, New York, N.Y., United States of America11 Johns Hopkins University, Baltimore, Maryland, United States of America12 Lawrence Livermore National Laboratory, Livermore, California, United States of America13 Euratom-UKAEA Fusion Association, Abingdon, Oxfordshire, United Kingdom14 Massachusetts Institute of Technology, Cambridge, Massachusetts, United States of America15 General Atomics, San Diego, California, United States of America16 Compx, Del Mar, California, United States of America17 University of California, Irvine, California, United States of America18 Princeton Scientific Instruments, Princeton, New Jersey, United States of America19 Nova Photonics, Princeton, New Jersey, United States of America20 Kyushu Tokai University, Kumamoto, Japan21 Himeji Institute of Technology, Okayama, Japan22 Hiroshima University, Hiroshima, Japan23 ENEA, Frascati, Italy24 University of Wisconsin, Madison, Wisconsin, United States of America25 Tokyo University, Tokyo, Japan26 Sandia National Laboratories, Albuquerque, New Mexico, United States of America27 Korea Basic Science Institute, Taejon, Republic of Korea

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Abstract - A major research goal of the National Spherical Torus Experiment is establishing long-pulse, highbeta, high confinement operation and its physics basis. This research has been enabled by facility capabilitiesdeveloped over the last two years, including neutral beam (up to 7 MW) and high harmonic fast wave heating(up to 6 MW), toroidal fields up to 6 kG, plasma currents up to 1.5 MA, flexible shape control, and wallpreparation techniques. These capabilities have enabled the generation of plasmas with <βT> up to 35%.Normalized beta values often exceed the no wall limit, and studies suggest that passive wall mode stabilizationis enabling this for broad pressure profiles characteristic of H mode plasmas. The viability of long, highbootstrap current fraction operations has been established for ELMing H mode plasmas with toroidal betavalues in excess of 15% and sustained for several current relaxation times. Improvements in wall conditioningand fueling are likely contributing to a reduction in H mode power thresholds. Electron thermal conduction isthe dominant thermal loss channel in auxiliary heated plasmas examined thus far. HHFW effectively heatselectrons, and its acceleration of fast beam ions has been observed. Evidence for HHFW current drive is bycomparing of the loop voltage evolution in plasmas with matched density and temperature profiles but varyingphases of launched HHFW waves. A peak heat flux of 10 MW/m2 has been measured in the H mode, withlarge asymmetries in the power deposition being observed between the inner and outer strike points. Non-inductive plasma startup studies have focused on coaxial helicity injection. With this technique, toroidalcurrents up to 400 kA have been driven, and studies to assess flux closure and coupling to other current drivetechniques have begun.

1. Introduction

With the advent of significant levels of auxiliary heating and maturing diagnostic andoperational capabilities over the last two years, the National Spherical Torus Experiment [1](NSTX) has begun intensive research aimed at establishing the physics basis for highperformance, long pulse, solenoid-free operations of the spherical torus [2] (ST) concept.This research is directed at developing an understanding of the physics of the ST operationalspace, developing tools to expand this space, and contributing broadly to toroidal science. Tothese ends, research in the last two years has focused on high beta MHD stability,confinement, high harmonic fast wave heating and current drive, boundary physics, solenoid-free startup, and exploration of scenarios that integrate favorable confinement, stability, andnon-inductive current drive properties. Some results in these efforts include the following:

• Toroidal beta values (βt ≡ <p>/(Bt02/2µ

0)) up to 35% with neutral beam heating have been

obtained. In some plasmas at high normalized beta βN ≡ βt/(Ip/aBt), the no-wall stabilitylimit is exceeded by 30%.

• Pulse lengths have been lengthened to 1 second with the benefit of bootstrap and beam-driven non-inductive currents of up to 60 % of the total.

• Normalized beta values βN up to 6.5 %·m·T/MA have been achieved, with operationsoverall bounded by ratios of βN to the internal inductance li = 10.

• Energy confinement times in plasmas with both L and H mode edges exceed theITER98pby(2) scaling [3] by over 50%, and the ITER89-P L-mode scaling [4] by over afactor of two for both discharge types.

• Particle transport studies of plasmas with turbulent (L mode) edge conditions revealimpurity transport rates that are consistent with and in some cases fall below neoclassicalpredictions in the core.

• Signatures of resistive wall modes have been observed [5,6]. With sufficiently broadpressure profiles, their onset occurs above the calculated no-wall stability limit, pointingto the presence of passive wall stabilization.

• Tearing mode activity consistent with the expected behavior of neoclassical tearing modeshas been observed. These modes can saturate beta or cause beta reduction when thecentral q value is near unity, but for higher q values their effect on performance is modest.

• New classes of fast-ion-induced MHD have been observed [7,8,9]. These CompressionalAlfvén eigenmodes (CAEs) exist near the ion cyclotron frequency. Bounce-precession

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Figure 1. Schematic cut-outview of the NSTX device.

Passive stabilizingplates

Center stack

Outer PF coils

Outer TF coils

Inner PFs and OHcoil

Inner TF coilsCeramic insulator

fishbone bursts are seen near 100 kHz, and are associated with fast ion losses.• Significant heating of electrons with high harmonic fast waves (HHFW) has been

measured [10]. Interactions between fast beam ions and HHFW have been observed.• The first indications of current driven by HHFW have been obtained [10].• The application of coaxial helicity injection [11,12] (CHI) has yielded a toroidal current

of up to 400 kA, with observations of n=1 MHD activity that may be a prerequisite forclosed flux surface formation.

• Edge heat flux studies [13] divertor infrared camera measurements indicate that 70% ofthe available power flows to the divertor targets in quiescent H mode discharges.

In this work, operational capability and diagnostics are described in Section 2. Section 3summarizes three studies aimed at realizing high toroidal beta, demonstrating long pulseoperations sustained by significant non-inductive current, and the combined realization ofhigh beta and efficient confinement for durations longer than an energy confinement time.Section 4 contains summaries of topical research in MHD, confinement, HHFW, boundaryphysics, and coaxial helicity injection. Particular attention is given to those elements relevantto establishing the physics and operational basis for long pulse, high beta, high confinementregimes with high fractions of non-inductive current drive.

2. NSTX Device Description and Facility Capabilities

Some of the NSTX device characteristics [14] and facilitycapabilities are as follows. NSTX can generate plasmas withan aspect ratio R/a as low as 1.27. Plasma currents up to 1.5MA have been obtained, and deuterium neutral beam injection(NBI) for heating and current drive is used routinely. Injectedin the direction of the plasma current, the NBI system iscapable of delivering 5 MW for up to 5 s. Powers up to 7 MWhave been achieved for shorter periods of time. High harmonicfast wave [15] can be delivered at variable phase for heatingand current drive. Injected powers up to 6 MW have beenachieved. NSTX has a close fitting conducting shell tomaximize the plasma beta. The toroidal field capability (BT ≤0.6 T) allows for pulse lengths up to 5 s at lower fields. Single

and double null configurations can be generated, andelongations up to 2.5 and triangularities up to 0.8 have beenachieved. Finally, the inner and outer halves of the vacuumvessel are electrically isolated from each other and can be biased for studies aimed at startingand sustaining the plasma non-inductively using coaxial helicity injection (CHI) [11].

Operational developments include development of 3500 C bakeout capability of the plasma-facing graphite tiles, implemented prior to the 2002 research campaign. This is part of a largerwall conditioning program [16] that includes routine application of helium glow betweenshots to reduce impurity influxes, as well as boronization every few weeks of operation or asdeemed necessary. Minimization of error fields by realignment of an outer poloidal field coillast year reduced the frequency of the onset of locked modes, widening the NSTX operatingspace. Finally, the capability of fueling the plasma from the center stack was implemented,motivated by work on the MAST device [17], complementing the outboard gas puffingcapability. One result of these improvements was improved access to and reproducibility of Hmodes.

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Figure 2. Surfaces of constantmagnetic flux for a neutral-beam heated plasma with atoroidal beta of 35% (shot108989). The plasma had anapplied toroidal field of 0.3 T,aspect ratio of 1.4, elongationof 2.0, triangularity of 0.8.

A schematic cross section of NSTX is shown in Figure 1 [18,19]. Central to NSTX researchis a suite of diagnostics [20], including a multitimepoint Thomson scattering system(presently 20 radial points, covering high field side to low field side, at up to 60 Hzsampling) that is absolutely calibrated for both density and temperature profilemeasurements. Carbon ion temperature and toroidalrotation measurements are made using charge exchangerecombination spectroscopy (CHERS), with a timeresolution of 20 ms (10 ms for the upcoming run period)and 17 radial channels (up to 50 channels for theupcoming run) spanning the outer half of the plasma crosssection. These measurements are facilitated by adedicated background view for direct subtraction ofcontaminating background. Ultra soft x-ray measurementsmade using three arrays, displaced toroidally andpoloidally, enable core MHD instabilities to be identified.An array of magnetic sensors on the center stack, as wellas the outboard side of the plasma, permit magneticequilibrium reconstruction and identification of toroidalnumber of external modes. A fast magnetic coil sensorsystem enables measurements of MHD perturbations atseveral times the ion cyclotron frequency (up to 10 MHz),in the range of compressional Alfvén eigenmodes [8]. Ascanning neutral particle analyzer measures the fast iondistribution function, including distortions induced by fastion absorption of high harmonic fast waves energy.Infrared cameras have enabled the first studies of edge heat flux scalings.

Key to progress in research on NSTX has been the development of a flexible shaping andposition control system. High triangularity and elongation raises the edge q for a fixedcurrent and toroidal field. Owing to the strong in-out variation of the toroidal field, theshaping can be particularly beneficial to low aspect ratio devices such as NSTX, includingthe realization of higher values of I/aB as compared to those achievable on larger aspect ratiodevices. Strong shaping also enables rapid current ramps, with typical ramp rates of up to 5MA/s utilized at the start of an NSTX pulse. Figure 2 shows the plasma equilibrium for theNSTX discharge with a toroidal beta of 35%. The equilibria are evaluated from magnetics-based equilibrium reconstructions using the EFIT code [21]. Recently, a control algorithmbased on real-time EFIT (rtEFIT) [22] reconstructions, originally developed at DIII-D, hasbeen implemented. This will enable improved position, shape, and feedback control in futureexperimental campaigns.

3. High beta, long pulse, and high confinement plasmas

3.1 High beta operations - The low aspect ratio and strong shaping capability on NSTX hasenabled the realization of high beta plasmas. Shown in Figure 3 are time traces from theplasma with the highest βT yet obtained. Run in the double null configuration at 0.3 T, thisplasma reached βT of 35%. The maximum βN was 6.3, below the no-wall limit. Thisdischarge entered a dithering H mode state near 230 ms, which transited to an ELM-freestate after 260 ms. Beta saturation was associated with the onset of an internal 1/1 mode.Depletion of available volt-seconds led to the termination of this discharge.

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3.2 Long pulses with significant non-inductivecurrent - One important goal for ST research in thelong term is the achievement of high fractions on non-inductive current drive. Progress towards this hasbeen realized through the generation of plasmas witha non-inductive current fraction of up to 60%,sustained for a duration on the order of a currentpenetration time (Figure 4). In these 0.45 T, 800 kAplasmas, β T = 15 – 20%. DCON [23] analysisindicates that this plasma, with βN of about 6, waswell above the no-wall stability limit, suggesting thatpassive wall stabilization played a role in itssustainment. Neutral beam current drive and bootstrapcurrents yielded low surface voltage of 0.1 V for atime period on the order of the estimated currentrelaxation time of about 200 ms. Calculations indicatethat late in the low loop voltage phase, more than halfof the non-inductive current comes from the bootstrapeffect. The total pulse lengths for these neutral-beam-heated H-mode plasmas extends to one second, with700 ms current flattops. Early neutral beam heating and the rapid current ramp combined toreduce the plasma’s internal inductance by yielding comparatively high edge currentdensities and slow current diffusion rates in the early phase of the discharge. Also, thedevelopment of an increased edge bootstrap current is calculated to be associated with the H

mode edge pedestal.

The pulse length was not limited by flux consumption. Rather,MHD activity at 650 ms that appears to be related to the qprofile and pressure profile evolution initiated the first drop incore beta. The details of this MHD are still under investigation.Measurements of core MHD activity with soft x-ray arrays areconsistent with the hypothesis that a double tearing mode thatfollows the generation of magnetic shear reversal is responsible

for the degradation. Confirmingthis awaits a direct measurement ofthe magnetic shear. As forincreasing the performance andpulse length of these plasmas,success in combining effectiveHHFW with neutral beam heatingcould have significant impact onthe plasma resistivity and bootstrapcurrent. Modification of the qprofile evolution with HHFW inthis manner, and ultimately with

HHFW aimed at direct current drive, represents a major research thrust for NSTX in theupcoming research campaign.

3.3 Simultaneous achievement of high stored energy and high confinement - Highertoroidal field operations led to the highest stored energies yet achieved in NSTX (Fig. 5), and

Figure 3. (a) Plasma current and NBpower, (b). Dα emission, and (c) βT fora double-null plasma. The plasma hadan applied toroidal field of 0.3 T,aspect ratio of 1.4, elongation of 2.0,triangularity of 0.8, internalinductance of 0.6, and a central q asdetermined from magnetics analysis of1 4

Figure 4. The plasmacurrent, injected beampower, surface voltage,beta poloidal, normalizedbeta, line density, and Dα

emission for an NSTXdischarge with over 50%non-inductive currentdr ive . The p lasmatransited to H modeshortly after the additionof the second neutral beam

108989EFIT02

Ip [MA]P NBI [MW]/10

βT [%]

Dα [a.u.]

0 0.1 0.2 0.3 0.4Time (s)

0

20

40

1

2

1

0.5

1.5

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the highest combined products of beta and confinement enhancement factor. These plasmashad an applied toroidal field of 0.55 T, higher than that used in the highest beta plasmas andnear the operational limit of 0.6 T. For this plasma, βNH89L, where H89L is the ratio of themeasured confinement time to that predicted by the ITER L mode scaling relation, is 12 orhigher for 8 energy confinement times, illustrating that high performance can be maintainedon NSTX for durations sufficient for the study ofthe physics of high beta, high confinementregimes.

4. Topical Research

4.1 MHD

4.1.1 Beta limiting modes - The simultaneousrealization of high values of normalized beta andlow internal inductance is one component ofdemonstrating the attractiveness and viability ofwall-stabilized, high bootstrap fraction operationsof the spherical torus [24]. Research on NSTX in2001 and 2002 has extended the range of βT, βN,βN/li, and pulse length achieved in a toroidalconfinement device of this scale. These plasmasstates have been achieved with confinement timesthat meet or exceed expectations based on scalinglaws developed from moderate aspect ratiotokamak experiments in both the L or H moderegime, as described in Section 4.2.

Some aspects of the operating space realized thus far on NSTX are illustrated in Figure 6. Atpresent, this operating space is bounded at βN/li = 10. One targeted operating point for NSTXplasmas is characterized by βN of 8 at li of 0.2 – 0.3 and is based on an assumption of broadpressure profiles so as to maximize the bootstrap current and stability. Achieving the highest

values of βN simultaneously withthe lowest values of internalinductance demands that NSTXoperate beyond the no-wall stabilitylimit. However, a challenge is tofind a path to these configurations.It is only for βN > 5 or more thatglobal MHD modes exhibitmagnetic perturbations withpoloidal wavelengths sufficientlylong for effective wall coupling. Todate, the path that has mostsuccessfully led to this corner of theNSTX operating space has utilizedrapid current ramps, early neutralbeam heating, shaping, andtransitions to the H mode that yield

broad pressure profiles. H mode pressure profiles also have benefits with respect to idealstability, bootstrap current generation, and large plasma volume with high energy content. It

Figure 6. The achieved values of (a) βN vs. plasmainternal inductance, and (b) βN vs. pressure peakingfactor for NSTX in the 1999 through 2002 researchcampaigns. The lighter points are data obtained in the2002 campaign. Beta and internal inductance isdetermined with EFIT code analysis of magnetics data.

UltimateGoal40% βT target

(a) (b)

Figure 5. Data for a plasma thatobtained 390 kJ of stored energy. An Hmode transition occurred at 300 ms. (a).Plasma current and neutral beamheating power. (b). Toroidal beta. (c).Energy confinement time and Dα

emission. (d). Stored energy, determinedfrom analysis of magnetics data.

Ip [MA]

PNBI [MW]/10

βT [%]

τE [s]

Dα [a.u.]

Wtot [kJ]

(a)

(b)

(c)

(d)

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should be pointed out, however, that plasmas with L mode edges in NSTX also exhibit highconfinement and toroidal beta values that approach or even exceed 30%.

Analysis of many plasmas with high βN indicate thatthe no-wall stability limit has been exceeded, and thatwall stabilization is likely a critical player in achievingthis state. Shown in Figure 7 are βN and the centralrotation for a discharge that exceeds the no-wallstability limit as calculated by the DCON code. Theno-wall limit is surpassed for several wall times.Significantly, the collapse of high plasma storedenergy and falling below the no-wall limit is precededby a reduction of the plasma rotation, suggesting thatthe wall mode stabilization that is enabled by thisrotation is lost at some critical rotation frequency. Adetailed analysis of passive wall stabilization in highβN NSTX plasmas is provided in Ref. [6].

Tearing mode activity, probably neoclassical tearingmodes, have been observed on NSTX to saturate beta insome cases, as well as to degrade overall performance.These modes are slowly growing and, in many shots, areidentified as 1/1, 2/1, and 3/2 islands. The mode growth isconsistent with that predicted by modified Rutherfordequation. These modes are most easily avoided byoperating plasmas with elevated q(0), which is consistent with the desired final highperformance state of low internal inductance operation.

A recent theoretical study [25] revealsthat the theoretical ideal beta limits ofmoderate aspect ratio tokamaks and ST’scan be viewed in a unified fashion if thestandard definition of beta is broadened.It is found that the normalized beta limitsover a wide range of aspect ratio aresimilar if the volume-averaged magneticfield pressure <β > ≡ 2µ 0<p>/<B2> isused to define <β N > ≡< β>(%)aBt0/Ip(MA), where Bt0 is theapplied vacuum magnetic field. Figure 6shows a database of normalized betavalues using both the usual definition and

this modified definition. The shaded band in Figure 8 shows the stability limit for moderateand small aspect ratio achieved in the theoretical study, along with the NSTX data. Theseplasmas exceed this theoretical limit, suggesting that some stabilization mechanism is atwork.

4.1.2 Fast beam-ion-induced MHD - In general, ST’s are susceptible to fast ion driveninstabilities due to the relatively low toroidal field. Indeed, a wide variety of suchinstabilities has been seen in NSTX at frequencies ranging from a few kHz to many MHz [9].In the frequency range below about 200 kHz, a form of the fishbone or energetic particle

Figure 8. (a) β T (black) and <β>. (red) atmaximum stored energy for NBI-heated plasmasplotted versus normalized current. Constant βN

lines are shown (b) <βN>. versus q* for thedischarges from (a). From Ref. 25.

Figure 7. βN and central carbonrotation velocity as a function oftime. This plasma exceeds the no-wall stability limit, as identified bythe DCON code, for several walltimes. The loss of stability ispreceded by a reduction of theplasma rotation.

%

tTime (s)

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mode has been seen, as well as modes that appear to be similar to the TAE modes ofconventional tokamaks. Unlike in conventional tokamaks, the frequency ranges of these two

classes of instabilities have substantial overlap,complicating the experimental identification andtheoretical analysis. Significant fast ion losses havebeen correlated, under some conditions, with theappearance of both of these types of modes.

In the higher frequency ranges, up to 5 MHz, orperhaps higher, observed modes may be related to thevarious forms of Ion Cyclotron Emission (ICE) ofconventional tokamaks. According to theoreticalmodeling, that both Global Alfvén Eigenmodes (GAE)and Compressional Alfvén Eigenmodes (CAE) couldbe destabilized by the fast ion population. There issome experimental/theoretical evidence that both havebeen observed on NSTX. Their presence could clearlyimpact fast ion distributions and thus, ultimately, fastion confinement, but no clear experimental evidenceexists for thisas yet.

An experiment has been performed in conjunctionwith the DIII-D tokamak that takes advantage of thesimilar cross-sectional shapes and area, but differentaspect ratios. Toroidal Alfvén Eigenmodes (TAE)modes were identified at similar frequencies in bothbut at higher mode numbers on DIII-D, as expectedby theory. Also, the threshold in beam beta for beam-driven instabilities is similar in both devices.

4.2 Confinement and transport

4.2.1 Global Confinement – The confinement timesin neutral beam heated NSTX plasmas comparefavorably to the ITER-89P empirical scalingexpression as well as the ITER-98(pby,2) scalingrule [26,27]. This is true for plasmas with distinct Hmode transitions as well as for L mode edge plasmas.An interesting aspect of this relation between the Hand L mode states can be seen in Figure 3. At the Hmode transition (near 230 ms), a change in plasmabeta and stored energy is not noticeable. While anincrease in the rate of change of stored energy isusually observed in L to H transitions, these othercases are prompting analysis of the local changes oftransport properties in the core and edge across an Lto H mode transition.

4.2.2 H mode access, dynamics, and power balance - H mode operations have becomeroutine on NSTX, aided by improved wall conditioning and reduced error fields. Access to

Figure 9. A sampling of theenergetic particle modes oftenobserved an NSTX neutral-beam-heated discharge.

Figure 10. Energy confinement times,determined with magnetics analysis,compared to values calculated from thetwo ITER scaling expressions. Data forplasmas with both L and H mode edgesare shown.

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the H mode is easiest in the lower single null configuration, but H modes have been obtainedin double null as well. The power threshold of several hundred kW in some cases and isexhibiting a secular fall as wall conditions improve [28].

The evolution of the density of the long-pulse H modeplasma described can be seen in Figure 11. Of note is thepresence of pronounced “ears” in the electron densityprofile that arise shortly after the L to H transition, likelya signature of an edge particle transport barrier.

A time slice of the profile measurements of the electrontemperature Te, the ion temperature Ti, and the rotationvelocity Vφ are shown in Figure 12. The high Ti

compared to Te is a persistent feature seen in mostNSTX neutral-beam-heated discharges. Along withexpectations that neutral beam fast ion energy should betransferred predominantly to the electrons in thistemperature range, this suggests that the dominant losschannel is electron thermal conduction. This is indeed born out in the power balance

analysis, which shows ion thermal conductivity χ i thatare on the order of predictions from neoclassical theory,and electron thermal conductivity χe that is significantlylarger than χi. The momentum diffusivity χφ is smallerthan χ i in this analysis, qualitatively consistent withexpectations from neoclassical theory. Analysis of neongas puffing also suggests that the impurity diffusion isnear neoclassical levels in the core of L mode plasmas.Microinstability analysis of beam-heated plasmas usingthe GS2 gyrokinetic code [29] is being performed toassess the roles of long wavelength ion temperaturegradient (ITG) and trapped electron (TEM) modes, aswell as shorter wavelength electron temperature gradient(ETG) modes. Further details of NSTX coreconfinement studies can be found in Ref. [30]

4.2.3 Edge turbulence measurements – Three different measurement techniques have beenimplemented to measure turbulence characteristics in the plasma scrapeoff layer. All of them– an edge reciprocating Langmuir probe, edge reflectometry, and gas puff imaging – revealhighly turbulent SOL activity in the L mode. The edge probe and imaging point to thepresence of intermittent convective transport events. These results will be compared todeveloping theory of these nonlinear transport phenomena. Studies will focus on assessingtheir role in determining the overall radial heat transport to the divertor.

4.3 High Harmonic Fast Wave Heating and Current Drive - Over the last two years, acampaign to explore the physics and the application of high harmonic fast wave (HHFW)heating has been carried out on the NSTX device with the ultimate goal of providing a tool forlong pulse, high beta ST operation. RF wave energy is launched into the NSTX plasma at afrequency of 30 MHz via a twelve-element antenna array. The elements can be phased tolaunch a variety of wave spectra with toroidal wave numbers between ±14 m-1.

As expected from theory, electron heating has been observed for a wide variety of plasma

Figure 11. Density evolution inthe long-pulse H mode describedin Figure 4.

Figure 12. Ti, Te, and toroidalrotation Vφ in the long pulse Hmode discharge described inFigure 4.

217 ms

233 ms

300 ms

400 ms

500 ms

600 ms

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conditions and over the full range of applied wave spectra. Electron temperatures as high as3.7 keV have been produced. Heating efficiency, characterized by the value of the centralelectron temperature, seems to be highest for the slowestwave phase velocities. NSTX plasmas withpredominantly electron heating exhibit a strongdegradation in confinement, which is consistent withtheoretical predictions of an increase in conduction dueto ETG modes. Attempts to measure the powerdeposition profile with modulated rf power have alsoshowed a very stiff temperature profile responseconsistent with a marginally stable profile. Somedischarges are characterized by an apparent barrier in theTe profile, and by reduced electron conduction in thecentral region. Long duration, 400 ms, steady rf drivenH-modes at moderate plasma current Ip =350 kA with βp

near unity, and bootstrap current fraction of 40% havebeen created. These H-modes have continuous ELM activity and are not accompanied by astrong increase in density or Zeff. H modes at higher values of plasma current have been ELM-free, with steadily increasing density until the termination after only a brief, ~40 ms interval.

Experiments with directed wave spectra have been conducted to investigate the possibility ofdriving plasma current. Differences between co and counter directed HHFW waves have been

observed in the loop voltage (Fig. 14) and in the MHDbehavior, which are consistent with current being driven onaxis in the expected direction. Careful matching of theelectron temperature and density are required to makemeaningful comparisons between the shots. For the best-matched cases at ± 7 m-1 a driven current of about 100 kAis inferred from the loop voltage, in reasonable agreementwith predictions from the TORIC code [31] and a factor oftwo smaller than that estimated from CURRAY [32]. Loopvoltage differences have also been observed for faster wavephase velocities down to ±2.4 m-1. An interesting and not

understood effect is that differences in central heatingefficiency are also found between co and counter phasingwith counter up to twice as efficient. Theory predicts nosubstantial difference in heating efficiency for theopposed spectra.

Ion heating provides a potential alternate channel for rfabsorption that can lower the efficiency of current drive.

Despite the large values of cyclotron harmonic involved (9-14), significant wave damping isexpected at large values of the ion beta. Acceleration of neutral beam injected 80 keV ions to140 keV has been observed. The ion tail is strongest at the highest values of toroidal field. Italso shows a small dependence on wave spectrum, decreasing for increasing phase velocity incontrast to theoretical predictions of the opposite behavior. No dependence on plasma currentwas observed and the observed dependence on injection energy shows a smaller interaction atlower voltage.

4.4 Boundary physics – Boundary physics research in NSTX focuses on power and particlebalance. High heat flux on the target plate has been measured in lower-single null (LSN)

Figure 14. Surface loop voltage fortwo HHFW-heated plasmas withco- and counter antenna phasing.The ne and Te profiles werematched by adjusting the heatingpower to be 2.2 MW co-phasing,1.2 MW counter phasing.

Figure 13. Electron temperature fora plasma heated with 3.5 MWHHFW over a 100 ms time interval.

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divertor plasmas. For example, the peak heat flux in a lower single null ELM-free H-modeplasmas with 4.5 MW of heating power has reached 10 MWm-2, with a full-width half-max of2 cm at the outer target plate [28], approaching the spatial resolution of the IR camera used tomake the measurement (Fig. 15). Peak heat flux in H-mode plasmas increases with NBIheating power. The peak heat flux at the inboard target is typically 0.5 – 1.5 MW/m2, with aprofile full-width half max. of ~ 10 cm. The powerflowing to the inboard side is typically 0.2 – 0.33of the outboard power (Fig. 15). Outer target tileheating and incident power appear to be higher inL-mode plasmas than ELM-free H-mode plasmas,whereas the inboard sides are comparable.

A tile temperature increase of 300 °C has beenmeasured during the first 0.2 s after divertorestablishment in H-modes; extrapolation of thetemperature rise, assuming an increase ~ (time)1/2

with constant peak heat flux, yields a tiletemperature in excess of the 1200 °C engineeringlimit after ~ 3 s. While this limitation should not impact the NSTX near-term program ofinvestigating pulse lengths up to several energy confinement times, more detailed study isrequired to assess the power handling requirements for pulse lengths in excess of severalcurrent penetration times (beyond 1 second) at the highest available input powers. As pulselengths are increased in NSTX, the emphasis in boundary physics research will be placed onusing double-nulls, radiative divertor/mantle solutions, and X-point sweeping.

4.5 Coaxial Helicity Injection - Strategies for initiating current non-inductively on NSTXby the process of coaxial helicity injection (CHI [11,12]) are being developed. CHI isimplemented on NSTX by driving current along field lines that connect the inner and outer

lower divertor plates. A description of the CHI systemon NSTX can be found in Reference [33].A 50kA, 1kVDC power supply is connected across the inner andouter vessel components, which are insulated from eachother by ceramic rings at the bottom and top. The CHImethod drives current initially on open field lines,creating a current density profile that is hollow andintrinsically unstable. Taylor relaxation predicts aflattening of this current profile through a process ofmagnetic reconnection leading to current being driventhroughout the volume, including closed field lines.

The applied injector voltage determines the amount ofinjector current that can be driven for each combinationof toroidal field, injector flux and gas pressure. This fluxis defined as the difference in poloidal flux between theupper and lower insulating gaps separating the inner andouter electrodes. In this discharge shown in Figure 16, asthe injector voltage is increased and the injector fluxreduced, the toroidal current reaches nearly 400kA. The

injector current is 28kA, which results in a current multiplication factor of 14, roughly equalto the theoretical maximum attainable. During the high current phase after 200 ms, there areoscillations in the toroidal current signal. It is not known if these are signatures of

Figure 16. Injector parameters,and toroidal and coil current ina CHI discharge

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Figure 15. Divertor heat flux in twoquiescent H modes, one with 4.1 MWinjected NB power, the other with 4.9 MW

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reconnection events that lead to closed flux plasma that then decays, only to be re-established.In high current discharges such as this, a magnetic perturbation with amplitude 2mT measuredat the outboard midplane and toroidal mode number n=1 is observed, rotating toroidally in theEr × Bp direction with a frequency in the range 5 – 12kHz. Such a mode, which has been seenon high current CHI discharges on HIT and HIT-II, may be a signature of the formation ofclosed flux surfaces.

Plans on NSTX are to produce higher current, longer discharges for characterization with andwithout auxiliary heating. To this end, the absorber region has been modified to suppress arcs,and feedback control should help to retain the closed flux that may be created in future highercurrent CHI discharges, a necessary step in performing a successful handoff to other currentdrive scenarios. A new short pulse discharge initiation method, developed on HIT-II, will beused on NSTX to hand-off a CHI discharge for inductive operation [34].

Summary: This research was supported by DoE contract DE-AC02-76CH03073. [1] M.Ono, Proceedings of the 18th IAEA Meeting, Sorrento, Italy, 2000.[2] Y-K M. Peng, and D. J. Strickler, Nuclear Fusion 26 (1986) 576[3] The ITER Team, Nucl. Fusion 39 (1999) 2137.[4] P.N. Yushmanov et. al., Nucl. Fusion 30 (1990) 1999.[5] S. Sabbagh et al., Nucl. Fusion 41 (2001)1601.[6] S. Sabbagh et al., paper EX/S2-2, this conference.[7] N.N. Gorelenkov and C.Z. Cheng, C.Z., Nucl. Fusion 35 (1995) 1743.[8] E.D. Fredrickson, N.N. Gorelenkov, C.Z. Cheng et al., Phys Rev. Lett. 87 (2001) 145001.[9] D. Gates, R. White, N. Gorelenkov, Phys. Rev. Lett 87 (2001) 205003.[10] P. M. Ryan, paper EX/P2-13, this conference.[11] T.R. Jarboe, et al., Phys. of Plasmas 5 (1998) 1807-14[12] T.R. Jarboe et al., Proceedings of the 17th IAEA Fusion Energy Conference, Yokohama, IAEA-CN69/PDP/02 (1998).[13] R. Maingi et al., in Proceedings of the 11th International Conference on Plasma Physics (Sidney,Australia, 2002), Plasma Phys. Control. Fusion (in press).[14] M. Ono et al., in Proc. 18th IEEE/NPSS Symp. On Fusion Enginnering, Albuquerque, 1999, IEEE,Piscataway, NJ (1999) 53.[15] M. Ono Phys. Plasmas 2 (1995) 4075.[16] H.W. Kugel et al., J. Nucl. Mater. 290-293 (2001) 1185.[17] A. Sykes, et al., in Plasma Phys. Control. Fusion (1994) (Proc. 15th Int. Conf. Seville, 1994) 1, IAEA,Vienna (1995) 719.[18] C. Neumeyer et al., in Proc. 18th IEEE/NPSS Symp. On Fusion Engineering, Albuquerque, 1999, IEEE,Piscataway, NJ (1999).[19] J. Chrzankowski, et al., ibid.[20] R. Kaita et al., in the Proceedings 28th International Conference on Plasma Science (ICOPS) / 13thInternational Pulsed Power Plasma Science (PPPS-2001), Las Vegas, Nevada, June 17 - 22, 2001.[21] L. Lao et al Nucl. Fusion 30 (1990) 1035[22] J.R. Ferron et al., Nucl. Fusion, 38 (1998) 1055[23] A. H. Glasser and M. S. Chance, Bull. Am. Phys. Soc. 42, 1848 (1997).[24] J. E. Menard, S. C. Jardin, S. M. Kaye et al., Nucl. Fusion 37 595 (1997)[25] J. Menard et al., submitted to Phys. Rev. Lett.[26] Kaye S. et al., Nucl. Fus. 37 (1997) 1303[27] The ITER Team, Nucl. Fus. 39 (1999) 2175.[28] R. Maingi et al., paper EX/C2-5, this conference.[29] M. Kotschenreuther et al., Comp. Phys. Comm. 88 (1995) 128.[30] B. LeBlanc et al., paper EX/C5-2, this conference.[31] M. Brambilla, Plasma Phys. Control. Fusion 41 (1999) 1.[32] T.K. Mau et al., Proc. of the 13th Top. Conf. on Applications of RF Power to Plasmas, Annapolis,Maryland (American Institute of Physics, Melville, 1999) 148.[33] R. Raman et al., Nucl. Fusion 41 (2001).1081[34] R. Raman et al., submitted to Phys. Rev. Lett.